This invention relates to refueling systems for a nuclear reactor and more particularly to a system for transferring fuel assemblies between a reactor core and a fuel storage area while the fuel assemblies remain completely submerged in a continuous body of coolant.
In nuclear reactors, particularly sodium cooled fast breeder reactors, a reactor vessel having an inlet and an outlet for the circulation therethrough of a coolant has a closure head disposed thereon in a fluid tight relationship. A core comprising fuel assemblies is supported within the reactor vessel. The fuel assemblies comprise a plurality of fuel rods that contain nuclear fuel. Under appropriate conditions, the fuel rods generate heat in a manner well known in the art. The coolant circulating through the reactor vessel passes in a heat transfer relationship with the fuel assemblies thereby cooling the fuel rods while carrying the heat away to generate electricity in a commonly known fashion. After a period of operation, the nuclear fuel in the fuel rods becomes depleted necessitating the replacement of the fuel assemblies with fresh ones; a process generally known as refueling of the reactor. While being replaced, decay heat is generated by the fuel rods of the fuel assembly.
During the refueling of sodium cooled fast breeder reactors, it is necessary to remove the decay heat from the irradiated fuel assemblies to prevent their reaching a temperature at which the fuel rod ruptures, because the rupture of the fuel rod results in the release of radioactive gases thus presenting health and contamination problems that are unacceptable. Therefore, it has become accepted practice to provide a means of dissipating the decay heat of an irradiated assembly by providing a system whereby the fuel assembly is kept submerged in a sodium coolant during transfer between the reactor and decay storage.
In most under sodium refueling systems, the means of keeping the fuel assembly submerged in sodium has been to place the fuel assembly in a sodium filled pot, a container with an open end, and then transfer the pot to the desired location. With this type of under sodium refueling system, an auxiliary gas or liquid metal cooling system for the pot of sodium is always necessary to meet an emergency condition such as failure of the transfer means. In addition, the ever present possibility of failure of the auxiliary system must be dealt with by providing a backup system. Typically, the cooling capability of the gas system is limited to about (10) KW because of technology limitations. In large commercial fast breeder reactor plants the decay heat from a spent fuel assembly, at the time after reactor shutdown when it is economical to do refueling, is as much as (60) KW or greater. An auxiliary cooling system capable of safely removing that amount of heat, is not presently compatible with the entire reactor plant economics and technology.
In the British Prototype Fast Reactor (PFR) refueling system, the irradiated fuel assembly is stored in a rotor inside the reactor vessel for an initial decay period, and then moved through an opening in the reactor head into a transfer machine located outside of the reactor vessel, an ex-vessel transfer machine. The machine moves above an opening in a transfer tunnel, couples onto a mechanism around the opening, and lowers the fuel assembly through the opening. Equipment within the tunnel moves the fuel assembly laterally in the tunnel to beneath a second opening in the top of the tunnel. A crane mounted machine moves the fuel assembly through a valve on the opening into a fuel handling cell and into further storage prior to partial disassembly for shipping to reprocessing. Aside from the cost problems, there are problems associated with sophisticated interlocks that must be provided to insure against release of fission products during refueling because of the danger to operating personnel. In addition, the valves and adapters associated with coupling the ex-vessel transfer machine to the mechanism around the opening in the reactor head and to the mechanism around the opening in the tunnel are large and expensive. This coupling and uncoupling process is also quite time consuming which increases the refueling time.
The British Commercial Fast Reactor (CFR) refueling system stores the irradiated fuel assembly in a rotor inside the reactor vessel for an initial decay period, then moves it through an opening in the reactor head to a gas cooled compartment and then through an opening in the compartment into a sodium filled compartment outside of containment. The closures in the openings of the compartments are expensive to install and maintain.
The French Phenix and Phenix 4 refueling systems move the irradiated fuel assembly in a pot through an opening in the reactor head into a gas cooled compartment, then through an opening in the compartment into a decay storage rotor outside of the containment housing the reactor. After a decay period the fuel assembly is moved from the rotor through an opening into a fuel handling cell. In this concept a major safety problem would result if the cooling system of the gas cooled compartment fails and the sodium filled pot containing the fuel assembly becomes stuck in the compartment, as would happen in a power failure. This would result in the fuel assembly overheating and possibly rupturing the fuel rods releasing contaminants.
The Russian liquid metal power reactor refueling systems, in principle, handle the fuel assemblies similarly to the French and British, and are subject to similar problems.
In U.S. Pat. No. 3,915,792, to M. Aubert et al, issued Oct. 28, 1975, there is described a hood for transferring fuel assemblies between two separate areas in a flask filled with a coolant while attempting to maintain a barrier against contamination of the atmosphere. Because the Aubert patent describes the transfer of a fuel assembly from the reactor coolant body into a flask filled with coolant and then lifts the flask above the reactor coolant body, the fuel assembly is not transferred under a continuous body of coolant. Under these circumstances the Aubert patent does not solve the problem of the flask becoming immobilized above the reactor coolant body thereby possibly overheating.
In copending application Ser. No. 660,744, filed Feb. 23, 1976, entitled "Nuclear Reactor Fuel Transfer System", by E. E. Wade, which is assigned to the present assignee, there is described a system capable of transferring fuel assemblies between a nuclear reactor core and a fuel transfer area while the fuel assemblies remain completely submerged in a continuous body of coolant. However, that system employs a device for lifting fuel assemblies which is disposed at an angle on a tilted rotatable plug of the closure head. While the invention described in copending application Ser. No. 660,744 is an acceptable alternative, design considerations may proscribe the use of a tilted rotatable plug while suggesting the use of a substitute such as a pivotable apparatus.
Another system for refueling a nuclear reactor while the fuel assemblies remain completely submerged in a continuous body of coolant is described in copending application Ser. No. 660,746, filed Feb. 23, 1976, entitled "Ex-vessel Nuclear Fuel Transfer System", by E. E. Wade, and assigned to the present assignee. That system describes the use of a pivotable apparatus associated with rotatable closure head plugs but that system does not utilize a fuel assembly as the link activating the pivoting mechanism.
U.S. Pat. No. 3,909,350, to G. Dupuy, issued Sept. 30, 1975, describes a handling device comprising two tubular arms having synchronized movements for loading and unloading fuel assemblies within a nuclear reactor. In the Dupuy patent it is the entire upper arm that provides the vertical movement for lifting or lowering the chosen fuel assembly. In addition, the Dupuy patent does not teach a method of refueling the reactor while the fuel assemblies remain completely submerged in a continuous body of coolant.